This invention relates generally to methods and apparatus for monitoring nuclear reactors, and more particularly to such methods and apparatus that employ parameters which are monitored exterior of the reactor core.
Conventional pressurized water reactors typically contain a reactive region and commonly referred to as the core, in which sustained fission reactions occur to generate heat. The core includes a plurality of elongated fuel rods comprising fissile material, positioned in assemblies and arranged in a prescribed geometry governed by the physics of the nuclear reaction. Neutrons bombarding the fissile material promote the fissionable reaction which, in turn, releases additional neutrons to maintain a sustained process. The heat generated in the core is carried away by a cooling medium which circulates among the fuel assemblies, and is conveyed to heat exchangers which in turn produce steam for the generation of electricity.
A neutron absorbing element is also included within the cooling medium in controlled variable concentrations to modify the reactivity, and thus the heat generated within the core as required. In addition, control rods are interspersed among the fuel assemblies, longitudinally movable axially within the core, to control the core's reactivity and thus its power output. There are three general types of control rods that are typically employed for various purposes. Full length rods, which extend in length to at least the axial height of the core, are normally employed for reactivity control. Part length control rods, which have an axial length substantially less than the height of the core, are normally used for axial power distribution control. In addition, reactor shut-down control rods are provided for ceasing the sustained fissionable reaction with the core and shutting down the reactor. Both part length and full length control rods are arranged to be incrementally movable into and out of the core to obtain the degree of control desired.
While the radial power distribution of the core is fairly predictable, due to the prescribed arrangement of fuel assemblies and the positioning of control rods which are symmetrically situated radially throughout the core, the axial power distribution can vary greatly during reactor operation. Core axial power distribution has created many problems throughout the history of reactor operation for many reasons. Normally, coolant flow through the fuel assemblies is directed from a lower portion of the core to the upper core regions, resulting in a temperature gradient axially along the core. Changes in the rate of the fissionable reaction, which is temperature dependent will thus vary the axial power distribution. Secondly, the axial variation in the power distribution varies the xenon axial distribution, which further accentuates the variations in power axially along the core. This can lead to a xenon-induced axial power distribution oscillation which can, late in core life, be unstable without corrective operator intervention. Thirdly, insertion of the control rods from the top of the core, without proper consideration of the past operating history of the reactor, can worsen the axial power peaking.
Power distribution monitoring is typically conducted within conventional nuclear power plants through in-core instrumentation systems as well as ex-core instrumentations. In-core instrumentation systems are generally comprised of movable miniature fission chambers which are designed to yield information pertaining to neutron flux distribution at selected locations in the reactor core, fuel assembly outlet thermocouples, and in some cases fixed miniature fission chambers. Such systems provide an accurate measure of the core relative power distribution, but provide no automatic protective function for the reactor. On the other hand, ex-core instrumentation systems are typically comprised of uncompensated, long ion chambers or power range detectors which are located in four vertical instrument wells outside the reactor vessel and symmetrically placed with respect to the core. Such power range detectors are calibrated to the in-core system and are used to provide automatic reactor protection against adverse power peaking. As noted above, protection against axial power peaking which may arise due to divergent axial xenon oscillations and transient axial xenon redistribution are a prime interest, especially in those nuclear power plants which employ larger cores.
Over the lifetime of a nclear power plant, changing fuel management schemes can result in significant changes in both the magnitude and distribution of neutron flux and, hence, neutron fluence throughout the reactor vessel beltline region. In order to accurately access the long-term effects of neutron irradiation on the properties of reactor vessel materials, such changes in radiation level must be well known.
Operating reactors often utilize a reactor vessel surveillance program which consists of from six to eight surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor vessel wall. Neutron dosimeters contained in such surveillance capsules provide measurement capability at a single location within the reactor geometry. Nevertheless, by themselves they cannot provide the gradient information that is required to evaluate the impact of fuel management schemes, such as the incorporation of low leakage loading pattern, which may result in radical changes in neutron flux distributions from cycle to cycle. Additional information, however, can be obtained by the use of supplementary passive neutron dosimeters installed in the reactor cavity annulus between the reactor vessel wall and the primary shield.
Prior art approaches which have utilized such supplementary passive neutron dosimeters have typically hung the dosimeters by stainless steel, nickel, or iron wires at various locations within the reactor cavity. Accurate placement of the dosimeters, however, was difficult at best since in most nuclear power plants the cavity between the reactor vessel and the primary biological shield is narrow and often largely inaccessible. Furthermore, during the operations of such nuclear power plants, expansion and contraction of the reactor vessel upon heat-up and cool-down, as well as heavy ventilating air currents which pass over the dosimeters, have a potentially adverse impact on the accuracy and repeatability of dosimeter placement. Mechanical vibrations exist as well. It is, therefore, readily apparent that a method and apparatus for accurately and repetitively placing supplementary passive neutron dosimeters within the reactor cavity would be desirable.
Commercial operators of nuclear power plants are also reluctant to employ such prior art approaches for the placement of supplementary passive neutron dosimeters due to the interference they create with refueling operations. During a typical refueling operation (i.e., the replacement of a reactor's fuel assemblies upon the exhaustion of their fuel) the head assembly of the reactor vessel must be removed in order to withdraw the spent fuel assemblies. However, in order to provide shielding of potentially dangerous radiation during refueling, the reactor cavity is sealed off in order that the space above the reactor vessel may be flooded with water. This is a accomplished most often in one of two ways. A first approach merely clamps a heavy steel plate over the reactor cavity, with the steel plate including gaskets on either side of the cavity. The second approach utilizes the steel plate of the first approach, but additionally employs an inflatable bladder which serves to further seal the reactor cavity at a top portion thereof. As can be appreciated, any reactor cavity dosimetry system which is to be used in such nuclear power plants must avoid the possibility of interfering with the sealing properties of the steel plate and/or bladder, as well as to prevent punctures to the inflated bladder which could lead to breakage of the seal.
In addition to the access problems presented by varying reactor cavity configurations, problematic placement of a reactor cavity dosimetry system is further aggravated by the necessity of protective clothing for those personnel changing the dosimeters. As is conventional, in order to work about a deactivated reactor, personnel must wear many layers of protective clothing and a full-face respirator which can lead to heat prostration under long periods of use. It would, therefore, be desirable in the design of a reactor cavity dosimetry system to provide one which is capable of rapid and remote deployment.